Sara Vichi

Characterization of 41Ar production in air at a PET cyclotron facility

In the production of Positron Emission Tomography (PET) nuclides at a medical cyclotron facility 41Ar (T1/2 = 109.34 m) is produced by the activation of air due to the neutron flux, according to the 40Ar(n, γ)41Ar reaction. In this work, we describe a relatively inexpensive and readily reproducible methodology of air sampling that can be used for quantification of 41Ar during the routine production of PET nuclides. We report the results of an extensive measurement campaign in the cyclotron bunker and in the ducts of the ventilation system, before and after final filtering of the extracted air. Air Samples were analyzed using a gamma-ray spectrometry system equipped with HPGe detector, with proper correction of the efficiency calibration to account for the samples density. The results of measurement were then used to evaluate the Total Effective Dose (TED) to the population living in the surrounding areas, due to routine emissions in the operation of the cyclotron. The average 41Ar saturation yield per one liter of air emitted in the environment resulted to be (0.044 ± 0.007) Bq/(μA ⋅ dm3). The maximum value of TED for the critical group of the population, even considering an overestimated workload, was less than 0.19 μSv/year, well below the level of radiological relevance.

Efficiency calibration of a portable CZT detector for nondestructive activation assessment of a cyclotron bunker

ABSTRACT During the operational life of a PET Cyclotron, the concrete walls of the cyclotron vault are activated by the secondary neutron flux interacting with rare earths and metals present in the concrete or in reinforcement bars. For this reason when considering dismantling of such accelerators, the amount of radioactive waste has to be evaluated in advance to identify any critical issues or possible countermeasures to be taken to define an optimum decommissioning strategy. The aim of this work is to define a non-destructive in situ measurement methodology for a preliminary activation assessment of a cyclotron bunker with no need for core drilling. A very compact, USB-powered, CdZnTe (CZT) detector for gamma-ray spectrometry was used for the activation assessment of the site of installation of a GE PETtrace (16.5 MeV) cyclotron, routinely used in the production of positron-emitting radionuclides. Because of the complexity of measurement geometry, the efficiency calibration of the detector was performed via Monte Carlo (MC) simulations. The detector was accurately modelled using FLUKA, including a 5 cm lead shielding set-up. The MC model of the detector was validated for a wide range of energies and different source geometries, showing discrepancies below 5% for all tested sources. The efficiency curve for wall activation measurements was calculated, allowing a quantitative evaluation of activity concentration.

Assessment of the neutron dose field around a biomedical cyclotron: FLUKA simulation and experimental measurements

In the planning of a new cyclotron facility, an accurate knowledge of the radiation field around the accelerator is fundamental for the design of shielding, the protection of workers, the general public and the environment. Monte Carlo simulations can be very useful in this process, and their use is constantly increasing. However, few data have been published so far as regards the proper validation of Monte Carlo simulation against experimental measurements, particularly in the energy range of biomedical cyclotrons. In this work a detailed model of an existing installation of a GE PETtrace 16.5MeV cyclotron was developed using FLUKA. An extensive measurement campaign of the neutron ambient dose equivalent H∗(10) in marked positions around the cyclotron was conducted using a neutron rem-counter probe and CR39 neutron detectors. Data from a previous measurement campaign performed by our group using TLDs were also re-evaluated. The FLUKA model was then validated by comparing the results of high-statistics simulations with experimental data. In 10 out of 12 measurement locations, FLUKA simulations were in agreement within uncertainties with all the three different sets of experimental data; in the remaining 2 positions, the agreement was with 2/3 of the measurements. Our work allows to quantitatively validate our FLUKA simulation setup and confirms that Monte Carlo technique can produce accurate results in the energy range of biomedical cyclotrons.

Determination of the activity meter calibration factor for Rhenium-188

ABSTRACT An innovative epidermal surface treatment based on 188Re has been recently developed for squamous cell carcinomas of the skin. The planning and delivery of the treatment requires an accurate knowledge of the source activity and thus proper calibration of activity meters. However, reference sources for calibration purposes are not always available, as in the case of short-lived radionuclides. The aim of this work is to determine the calibration factors for 188Re by comparison of measurements with an independently calibrated HPGe spectrometer. Calibration factors were experimentally determined for two different activity meters, a Capintec CRC15 and a MecMurphil MP-DC. This study was conducted on a Rhenium-188 compound produced by OncoBeta® GmbH for Rhenium-SCT® therapy. The final calibration factors, relative to 137Cs, for the MecMurphil MP-DC were: 4.35 ± 0.06 for point-like sources and 4.48 ± 0.09 for 5 mL solutions. While the final calibration factors for the Capintec CRC15 were: 4.12 ± 0.07 for point-like sources and 4.60 ± 0.10 for 5 mL solutions. With the presented method, we managed to determine calibration factors with an uncertainty below 3%.

Production of Ga-68 with a General Electric PETtrace cyclotron by liquid target

PURPOSE In recent years the use of 68Ga (t1/2 = 67.84 min, β+: 88.88%) for the labelling of different PET radiopharmaceuticals has significantly increased. This work aims to evaluate the feasibility of the production of 68Ga via the 68Zn(p,n)68Ga reaction by proton irradiation of an enriched zinc solution, using a biomedical cyclotron, in order to satisfy its increasing demand. METHODS Irradiations of 1.7 Msolution of 68Zn(NO3)2 in 0.2 N HNO3 were conducted with a GE PETtrace cyclotron using a slightly modified version of the liquid target used for the production of fluorine-18. The proton beam energy was degraded to 12 MeV, in order to minimize the production of 67Ga through the68Zn(p,2n)67Ga reaction. The products activity was measured using a calibrated activity meter and a High Purity Germanium gamma-ray detector. RESULTS The saturation yield of68Ga amounts to (330 ± 20) MBq/µA, corresponding to a produced activity of68Ga at the EOB of (4.3 ± 0.3) GBq in a typical production run at 46 µA for 32 min. The radionuclidic purity of the68Ga in the final product, after the separation, is within the limits of the European Pharmacopoeia (>99.9%) up to 3 h after the EOB. Radiochemical separation up to a yield not lower than 75% was obtained using an automated purification module. The enriched material recovery efficiency resulted higher than 80-90%. CONCLUSIONS In summary, this approach provides clinically relevant amounts of68Ga by cyclotron irradiation of a liquid target, as a competitive alternative to the current production through the68Ge/68Ga generators.

Modeling of a Cyclotron Target for the Production of 11C with Geant4

BACKGROUND In medical cyclotron facilities, 11C is produced according to the 14N(p,α)11C reaction and widely employed in studies of prostate and brain cancers by Positron Emission Tomography. It is known from literature that the 11C-target assembly shows a reduction in efficiency during time, meaning a decrease of activity produced at the end of bombardment. This effect might depend on aspects which are still not completely known. OBJECTIVE Possible causes of the loss of performance of the 11C-target assembly were addressed by Monte Carlo simulations. METHODS Geant4 was used to model the 11C-target assembly of a GE PETtrace cyclotron. The physical and transport parameters to be used in the energy range of medical applications were extracted from literature data and 11C routine productions. The Monte Carlo assessment of 11C saturation yield was performed varying several parameters such as the proton energy and the angle of the target assembly with respect to the proton beam. RESULTS The estimated 11C saturation yield is in agreement with IAEA data at the energy of interest, while it is about 35% greater than the experimental value. A more comprehensive modeling of the target system, including thermodynamic effect, is required. The energy absorbed in the inner layer of the target chamber was up to 46.5 J/mm2 under typical irradiation conditions. CONCLUSION This study shows that Geant4 is potentially a useful tool to design and optimize targetry for PET radionuclide productions. Tests to choose the Geant4 physics libraries should be performed before using this tool with different energies and materials.

Radiation Protection Studies for Medical Particle Accelerators using Fluka Monte Carlo Code

Radiation protection (RP) in the use of medical cyclotrons involves many aspects both in the routine use and for the decommissioning of a site. Guidelines for site planning and installation, as well as for RP assessment, are given in international documents; however, the latter typically offer analytic methods of calculation of shielding and materials activation, in approximate or idealised geometry set-ups. The availability of Monte Carlo (MC) codes with accurate up-to-date libraries for transport and interaction of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of modern computers, makes the systematic use of simulations with realistic geometries possible, yielding equipment and site-specific evaluation of the source terms, shielding requirements and all quantities relevant to RP at the same time. In this work, the well-known FLUKA MC code was used to simulate different aspects of RP in the use of biomedical accelerators, particularly for the production of medical radioisotopes. In the context of the Young Professionals Award, held at the IRPA 14 conference, only a part of the complete work is presented. In particular, the simulation of the GE PETtrace cyclotron (16.5 MeV) installed at S. Orsola-Malpighi University Hospital evaluated the effective dose distribution around the equipment; the effective number of neutrons produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of 41Ar. The simulations were validated, in terms of physical and transport parameters to be used at the energy range of interest, through an extensive measurement campaign of the neutron environmental dose equivalent using a rem-counter and TLD dosemeters. The validated model was then used in the design and the licensing request of a new Positron Emission Tomography facility.

An innovative gamma-ray spectrometry system using a compact and portable CZT detector for radionuclidic purity tests of PET radiopharmaceuticals

ABSTRACT This work was aimed at characterizing and validating a very compact, USB-powered, CdZnTe detector for gamma-ray spectrometry applications in radiopharmacy. A GR1 model (by KromekTM, Sedgefield, UK) was utilized. The detector was calibrated in energy and efficiency. Samples of [18F]FDG and [68Ga]-DOTANOC were measured to assess the detector’s suitability in radionuclidic purity measurements. The energy resolution and the efficiency obtained are fully adequate for quality control of positron emission tomography radiopharmaceuticals. Results of [18F]FDG and [68Ga]-DOTANOC samples proved to be comparable to those obtained using high-purity Germanium detectors. The performance of the detector and its very compact size make this type of device an extremely attractive tool for QC applications in radiopharmacy.

Radiation protection issues in the excavation of road and railway tunnels, a preliminary assessment

ABSTRACT In road and railway tunnel excavation, personnel are exposed to natural radiation. Exposure can have varying degrees of significance, as naturally occurring radioactive material (NORM) content may vary by more than one order of magnitude depending on local lithology and, as far as radon is concerned, depending also on features of more distant bedrocks involved in radon convection. Radon exhalation and build up in the confined environment of the tunnel is affected by permeability of rocks, which is highly variable with lithology. Excavation methods as well have an impact on workers’ exposure, due both to exposure time and to protection afforded by the equipment. Therefore, it is of interest to develop a method to predict radon levels to characterize the work area. Since drill cores are invariably taken before excavation begins, for a preliminary assessment of the geology involved, a method is proposed here for predicting radon levels, based on NORM and exhalation measurements in drill core samples, and on the geometry of the tunnel to be excavated. An experimental campaign is under way on a railway tunnel under construction in northwestern Italy: preliminary results will be presented and predicted values compared to experimental measurements.